Conference Agenda

A generic Small Modular Reactor (SMR) simulation model was developed in two system codes: Apros and TRACE. NuScale design data and other public SMR design data was used as a reference point for the development of the model. The objective of the work was to study the modelling choices and simulation capabilities of the selected codes with respect to the SMR design features (e.g. natural circulation systems, helical coil steam generator, compact vacuum containment, reactor pool and integrated pressurizer). In particular, the goal was to assess the suitability of Apros and TRACE simulation codes for the simulation of SMR applications. This was done by calculating one steady-state and three transient simulations (inadvertent operation of emergency core cooling system and two variations of turbine trip) with the developed simulation models. The results were compared with the reference simulation results presented in NuScale final safety analysis report (FSAR) to assess the capability of the codes and suitability of the modelling choices. One of the turbine trip cases was also compared with two previously published reference results by two other codes to complement the comparison and provide insight into the analysis of the results. Good match with the overall trend of the reference results was achieved with both Apros and TRACE simulation models which confirms the capability of the codes to model this type of SMR configuration and simulate both steady-state and typical transients.

This study experimentally investigates the frictional pressure loss and flow regime behavior of a 9×9 mock Helical Cruciform Fuel (HCF) rod bundle, a novel design proposed as a potential replacement for conventional cylindrical rods in Light Water Reactors (LWRs). The unique cruciform cross-section, featuring four twisted petals, eliminates the need for conventional spacer grids, offering higher fuel packing fraction and enhanced coolant mixing. To assess these advantages, a high-precision differential pressure measurement system was employed over a Reynolds number range of 200 to 22,000, covering laminar, transition, and turbulent flow regimes. The experimentally determined friction factors showed statistically similar trends between the “one pitch” and “bundle-averaged” axial segments, confirming fully developed flow in both regions. Empirical correlations for friction factor and differential pressure per unit length were then developed for each flow regime and validated by comparison to previous HCF and wire-wrapped fuel bundle studies. Results identified flow regime boundaries at approximately Re ≈ 1,000 for laminar-to-transition and Re ≈ 8,274 for transition-to-turbulent, highlighting distinctly different hydraulic behavior in the three regimes. The findings significantly broaden the limited experimental database on HCF rod bundles, providing new insights into regime-dependent pressure drop characteristics. By refining existing correlations and offering high-fidelity benchmark data, this work advances the development of more efficient and accurate reactor core designs that leverage HCF technology for enhanced thermal performance.

NuScale is a light water cooled small modular reactor with an integral reactor pressure vessel design that relies on natural circulation to provide the primary mass flow. This work focuses on the simulation of LOCA sequences caused by a break in the CVCS discharge line inside the steel containment. For this purpose, a model of NuScale was developed using the TRACE system code, which includes modeling of the primary and secondary systems, the steel containment, the reactor pool, and the safety systems. In this study, the base case corresponds to a LOCA in which the ECCS fails without opening either of the reactor recirculation valves. This scenario is selected based on the PRA results included in the NuScale DCA. A sensitivity analysis is then performed to determine the time available to manually actuate CVCS injection. Further simulations were also performed with the recovery of one out of two RRV openings. The results allow comparison of the time available for each LOCA management strategy.

In recent years, development of novel passive safety systems for new reactor designs has significantly increased. These systems are recognized for their ability to operate reliably for extended period of time and without the need for operator action or active components requiring electricity. The Innovative Small Modular Reactor (i-SMR), an integral-type SMR that is currently being developed in South Korea, incorporates two such passive safety systems. The Passive Auxiliary Feedwater System (PAFS) is intended for long-term core cooling and decay heat removal by condensation of steam removed from the steam generator. The Passive Containment Cooling System (PCCS) is designed to depressurize the containment vessel during a Loss of Coolant Accident (LOCA), replacing the conventional Containment Spray System (CSS). The performance of both PAFS and PCCS is governed by a heat transfer driven by a two-phase natural circulation flow, presenting several design challenges. Traditional deterministic safety assessment using system codes lack the precision needed to capture the detailed dynamics of phenomena occurring within passive safety systems, such as rapid steam condensation and associated multi-dimensional flow. Accurate prediction of the PAFS and PCCS performance under accident conditions necessitates a thorough understanding of these behaviors. This study therefore leverages the MULTID component for reliable simulation of the dynamic three-dimensional phenomena associated with operation of the passive safety systems, along with the overall plant response, evaluated using the MARS-KS. The main focus of this study is the EDV-LOCA scenario, where both PAFS and PCCS play a crucial role for effective accident mitigation.

Westinghouse AP300^TM SMR is the latest Westinghouse small modular reactor based on the proven AP1000® pressurized water reactor technologies to accelerate its development and deployment. The passive safety system of AP300 is the same but scaled down from the industry leading AP1000 passive safety system, which has been extensively analyzed and tested. The testing basis of AP300 is expected to be well covered by the extensive AP600/AP1000 testing programs, which consists of many separate effects test facilities and integral effects facilities for both passive core cooling system and passive containment cooling system, such as APEX600/1000, ROSA-AP, SPES, Madison CMT test, VAPORE, PRHR HX test, LST, PCS water distribution test, condensation test, etc. In addition, the program also includes the previous large break LOCA experiments that are essential for the licensing of AP600/AP1000 such as UPTF and CCTF experiments. These experiments will be reviewed and the applicability of the facilities and the experiments to the AP300 SMR will be discussed.